Abstract:

An integral pressurized water nuclear reactor for the production of steam
utilizing a helical coil steam generator, a plurality of internal
circulation pumps, and an internal control rod drive mechanism structure.

Claims:

1. A nuclear reactor comprising:a pressure vessel;a reactor core disposed
in the pressure vessel;a shroud disposed in the pressure vessel and
arranged respective to the reactor core such that primary coolant
circulates inside and outside the shroud; anda helical steam generator
disposed in the upper section of the pressure vessel and
circumferentially wrapped around the shroud.

2. The nuclear reactor as set forth in claim 2, further comprising at
least one decay heat removal system heat exchanger disposed in the
pressure vessel at a position in a lower vessel section of the pressure
vessel and relatively closer to the reactor core than the helical steam
generator.

3. The nuclear reactor as set forth in claim 2, wherein the at least one
decay heat removal system heat exchanger comprises a plurality of said
heat exchangers.

4. The nuclear reactor as set forth in claim 3, wherein the pressure
vessel has to reactor core disposed at or near a bottom of the pressure
vessel, and has no pressure vessel penetrations whose failure is capable
of generating a loss of coolant accident located at or below a level of
the at least one decay heat removal system heat exchanger.

5. The nuclear reactor as set forth in claim 4, wherein the at least one
decay heat removal system heat exchanger is designed to operate in a
condensation mode.

6. The nuclear reactor as set forth in claims 5, wherein the reactor core
has a polygonal cross section and the at least one decay heat removal
system heat exchanger is disposed in a space defined by a wall of a lower
shroud a lower portion of the pressure vessel.

7. The nuclear reactor as set forth in claims 5, wherein the reactor core
has a circular cross section and the at least one decay heat removal
system heat exchanger is disposed in a space defined by a wall of a lower
shroud a lower portion of the pressure vessel.

9. The nuclear reactor as set forth in claim 8, wherein the at least one
primary coolant circulating pump is not connected with a steam generator.

10. The nuclear reactor as set forth in claims 9, further comprising:an
internal steam generator disposed in the pressure vessel, the at least
one primary coolant circulating pump being disposed in the pressure
vessel between the internal steam generator and the reactor core.

11. The nuclear reactor as set forth in claim 10, wherein the pressure
vessel includes a mid-flange disposed between the internal steam
generator and the reactor core that supports the at least one primary
coolant circulating pump.

12. The nuclear reactor as set forth in claim 11, wherein the pressure
vessel has three sections including a lower vessel section housing the
reactor core, an upper vessel section housing the internal steam
generator, and an upper internals including a control rod drive mechanism
disposed between the internal steam generator and the reactor core, the
at least one primary coolant circulating pump being disposed proximate to
the upper internals.

13. A nuclear reactor comprising:a pressure vessel including a lower
vessel section and an upper vessel section;a reactor core disposed in the
pressure vessel in the lower vessel section;an internal steam generator
disposed in the pressure vessel in the upper vessel section; andupper
internals including at least one wholly internal control rod drive
mechanism.

14. The nuclear reactor as set forth in claim 13, wherein the pressure
vessel further includes a mid-flange at the junction between the lower
vessel section and the upper vessel section, the mid-flange mechanically
supporting the upper internals.

15. The nuclear reactor as set forth in claim 14, wherein the pressure
vessel further includes as shroud, having an inner and an out
circumference, positioned above the reactor core and arranged respective
to the reactor core such that primary coolant circulates upward within
the inner circumference of the shroud and downward between the outer
circumference of the shroud and the pressure vessel.

16. The nuclear reactor as set forth in claim 15, wherein the internal
steam generator is a helical coil steam generator comprising of a
plurality of steam tubes helically wrapped around an upper portion of the
shroud at helix angles between about 4 and about 10 degrees.

17. The nuclear reactor as set forth in claim 16, wherein the helical coil
steam generator includes at least two intertwined tube bundles, each
bundle have an independent tube sheet.

18. The nuclear reactor as set forth in claim 17, wherein the upper
internals include a plurality of primary coolant pumps located along the
outer circumference of the shroud, internal to the pressure vessel, and
below the helical coil steam generator.

19. The nuclear reactor as set forth in claim 18, wherein the pressure
vessel further includes at least one decay heat removal heat exchanges
located in a lower portion of the pressure vessel and below the plurality
of coolant pumps.

20. The nuclear reactor as set forth in claim 19, wherein the pressure
vessel is self pressurized via a steam bubble located above the shroud in
an upper portion of the pressure vessel.

[0002]The following relates to the nuclear power arts, nuclear power
safety arts, nuclear reactor control arts, and related arts.

[0003]Existing integral pressurized water reactor (PWR) designs place the
core generally at the bottom, with steam generators overlapping or wholly
above the core in the vertical direction. These PWR design can employ
natural convection where the core heats the primary coolant which then
rises in a central riser and then comes back down in an outer annular
region, sometimes called the downcomer region. Alternatively, forced
convection can be employed, in which the circulation of the primary
coolant is driven by pumps. In the latter design, the pumps are disposed
at the bottom of the reactor or above the steam generators in the
downcomer region and coupled with the steam generators to force the
primary coolant downward into the steam generators.

[0004]Some examples of such reactors are disclosed in: U.S. Pat. No.
4,072,563; U.S. Pat. No. 6,813,328; "Consolidated Nuclear Steam Generator
for Marine Application", The Engineer (Aug. 21, 1964); and U.S. Pat. No.
4,072,563, each of which is incorporated herein by reference in its
entirety.

BRIEF SUMMARY

[0005]In one aspect, a PWR having a three-section design is disclosed. The
reactor core is located in the lower section, the steam generators are
located wholly above the core in an upper section, and the pumps are
placed in the downcomer region at a third middle section located in the
gap between the upper section with the steam generators and the lower
section with the core. In this middle location, the pumps are optimally
positioned downstream of coolant flow such that that the pump operate in
a relatively cooler environment. Further, more even circulation is
achieved, as the pumps both "pull" primary coolant through the steam
generators and "push" primary coolant downward to the core. An axial flow
pump(s) may be utilized.

[0006]Another aspect of the disclosure is that all reactor vessel
penetrations that can result in significant reactor coolant loss during
operation are located far above the core.

[0007]In another aspect a gap between the steam generators and the core is
also available as a mechanical support for the control rod drive
mechanisms (CRDM) and related structures.

[0008]In another aspect of the disclosure, the steam generator tubes are
helically wrapped around the primary flow riser pipe. Packing efficiency
can be very high and heat is transferred from the downward flowing
primary coolant and possibly also from the primary coolant flowing upward
through the primary flow riser pipe, thus reducing opportunity for heat
loss and improving efficiency.

[0009]In another aspect of the disclosure, dedicated passive decay heat
removal system (DHRS) heat exchangers are provided, and are located
proximate to the reactor core. This places the heat exchangers in optimal
position for passive decay heat removal, and eliminates the complex
valving involved with using the steam generators as components of the
DHRS since the disclosed dedicated passive DHRS is wholly separate from
the steam generator. Redundancy can be incorporated by providing several
DHRS heat exchangers. Optionally, the steam generators may also be
connected with the emergency cooling tanks for further redundancy, with
this latter connection valved off in the event primary coolant level
drops low enough to expose the steam generators. Since the dedicated DHRS
heat exchangers are near the reactor core, they remain operative so long
as the primary coolant is at least high enough to immerse the core. These
heat exchangers may also be designed to operate in a condensation mode,
flowing primary coolant that has turned to steam.

[0010]In another aspect of the disclosure, a method of supporting the core
including individual fuel assemblies in a modular reactor is disclosed.
The fuel assemblies are supported by a bottom grid structure that is part
of a welded core support structure of 304L stainless steel or another
suitable material which is suspended from a flange on the lower reactor
vessel. Lugs on the inside of the lower reactor vessel center the core
support structure. The fuel assembly is supported laterally within the
core support structure via a core former assembly. This assembly is
optionally octagonal in shape to allow for transfer of the entire set of
fuel assemblies to the spent fuel pool. The optional octagonal shape also
provides space for dedicated decay heat removal heat exchangers to be
disposed between the lower vertical support shroud and the lower reactor
vessel. The core former assembly is suitably a welded and machined
structure made of 304L stainless steel or another suitable material. The
lower shroud assembly supports the fuel assemblies and the core former
structure, and is suitably a welded and machined structure of 304L
stainless steel or another suitable material. The upper portion of the
lower shroud optionally contains bypass flow orifices which allow the
decay heat removal heat exchangers to function when the water level is
too low to circulate past the steam generator, such as may be the case
during core refueling or certain types of loss of coolant accident (LOCA)
conditions. The bypass flow orifices are optionally located in the core
support shroud in an area where the outside diameter is increased within
the lower reactor vessel to provide a reduced area hence higher flow
velocity in the annulus between the reactor vessel and the core support
shroud during normal operation to prevent reactor coolant from flowing
inward through the bypass orifices and bypassing the core. The disclosed
design of the core support assembly provides for complete core refueling.

[0011]In yet another aspect of the disclosure, the above aspects are
combined to form a new and unique PWR design.

BRIEF DESCRIPTION OF THE DRAWINGS

[0012]The invention may take form in various components and arrangements
of components, and in various process operations and arrangements of
process operations. The drawings are only for purposes of illustrating
preferred embodiments and are not to be construed as limiting the
invention.

[0013]FIG. 1 is a sectional side view of a pressurized water reactor
(PWR).

[0027]FIG. 14 is a perspective view of a PWR upper internals pump
assembly.

DETAILED DESCRIPTION OF PREFERRED EMBODIMENTS

[0028]Referring to the drawing generally, wherein like reference numerals
designate same of functionally similar parts, FIG. 1 is a section side
view of a pressurized water reactor (PWR), generally designated 10,
according to the present invention. The illustrated primary vessel
contains the reactor core 50, internal helical steam generators 20, and
internal control rod drive mechanisms 40 (CRDM). The illustrative reactor
vessel includes four major components, namely: the lower vessel 1, the
upper internals 2, the upper vessel 3, and the upper vessel head 4. Other
vessel configurations are also contemplated.

[0029]With continuing reference to FIG. 1 and with further reference to
FIG. 2 which shows the lower vessel 1, the lower vessel 1 contains the
reactor core 50. The reactor core 50 can have substantially any
configuration compatible with a light water reactor. In one suitable
embodiment, the reactor core 50 includes 69 shortened 17×17 PWR
type fuel assemblies supported by a bottom grid structure that is part of
a core former structure 51. An illustration of a core former structure 51
is shown in FIG. 2A.

[0030]The illustrated octagonal configuration allows for transfer of the
entire set of fuel assemblies 550 to the spent fuel pool. The illustrated
octagonal shape of the lower vertical support shroud 52 (FIG. 2B) also
provides space 56 for dedicated decay heat removal system (DHRS) heat
exchangers 55 to be disposed between the lower vertical support shroud 52
and the lower reactor vessel 1 (FIG. 2).

[0031]The core former structure 51 is suitably a welded and machined
structure made of 304L stainless steel or another suitable material. The
lower shroud assembly 52 (FIG. 2B) supports the core former 51 and fuel
assemblies 55. The upper portion of the lower shroud 52 optionally
contains bypass flow orifices 67 which allow the DHRS heat exchangers to
function when the water level is too low to circulate past the steam
generator, such as may be the case during core refueling or certain types
of loss of coolant accident (LOCA) conditions. The bypass flow orifices
67 are optionally located in the lower support shroud 52 in an area where
the outside diameter is increased within the lower reactor vessel to
provide a reduced area hence higher flow velocity in the annulus between
the reactor vessel and the core support shroud during normal operation to
prevent reactor coolant from flowing inward through the bypass orifices
and bypassing the core.

[0032]Optionally, this core former structure 51 is replaced with each
refueling. The core former 51 is supported by a lower shroud 52 which is
in turn, is supported from a flange 53 of the upper spool portion 6 of
the lower vessel 1 and is centered by lugs attached to the inside of the
lower vessel elliptical head. Optionally, these lugs can also support the
weight of the core 50 in the event that the lower shroud 52 should need a
structural supporting member. An illustration of a lower shroud 52 is
shown in FIG. 2B.

[0033]The lower vessel 1 may house a dedicated decay heat removal system
(DHRS) in an annulus between the lower shroud 52 and the lower vessel 1.
In the illustration of FIGS. 3 and 3A, redundant DHRS heat exchangers 55,
each of which is a helical coil design (straight tube designs are also
contemplated) with primary coolant on the shell side, are shown. The
tubes are connected to tube sheets 54 attached to the nozzles 57. Tube
size, thickness and material is optionally the same as the steam
generators discussed later herein. The dedicated decay heat removal
system heat exchangers 55 can be disposed elsewhere in proximity to the
reactor core 50, with the location preferably selected to optimize their
utility as components of a DHRS. Advantageously, because these heat
exchangers 55 are not part of the steam generation system they can be
placed optimally for decay heat removal in the event of a reactor
emergency. For example, by extending the dedicated heat exchangers 55 of
the DHRS low within the pressure vessel, they remain useful even during a
loss of coolant accident (LOCA) in which the primary coolant level drops
below the level of the steam generators.

[0034]With further reference to FIG. 2, in some embodiments the lower
vessel 1 comprises three forgings of SA508 Gr 4N Cl 2 carbon steel. One
forging is the lower head 1A which may be of a circular or elliptical
configuration and in one configuration is about 4.125'' thick. The second
forging is the cylindrical shell 1B section which may be about 4.75''
thick and the third forging is the upper spool/flange section 1C. The
final weldment is preferably clad internally with 0.25'' stainless steel.
This illustrative arrangement of the vessel sections has the advantage of
eliminating welds near the core mid-plane where the fluence levels are
highest.

[0035]With reference to FIG. 1 and with further reference to FIG. 4, the
upper internals structure 2 provides support for control rod drives 40
and control rod guide frames 44 and is also a suitable structure through
which control rod drive power and control instrumentation signals can
pass. This allows the upper vessel 3 and integral steam generator 20 to
be removed independently of the control rod drives 40 and associated
structure 48. The upper internals structure 2 is generally composed of an
upper internals basket 46, a CRDM support structure 48, control rod guide
frames 44, a mid flange 29, and the control rod drive mechanisms 40.

[0036]With continuing reference to FIG. 1 and with further reference to
FIG. 5, the upper vessel 3 houses the steam generators 20, provides the
connections to the feedwater inlet 21 and steam outlet 22 lines and may
include penetrations for the reactor coolant inventory and purification
system (not shown). The steam generator 20 includes a cylindrical inner
shell 24 which is the upper flow shroud 24 which structurally separates
the central riser region 90 from the annular down-corner region 92 in
which the helical steam generators are located.

[0037]With further reference to FIGS. 6A through 6D, the steam generator
20 is a helical coil tube design which in one contemplated embodiment has
an inner diameter (ID) of about 66 inches and an outer diameter (OD) of
about 120 inches. Other dimensions are also contemplated. The steam
generator 20 optionally includes a plurality of (e.g., four, six, or
eight) separate, intertwined sections which permits isolation of any
section so the plant can continue to operate at reduced power. The tubes
are connected at each end by pigtail sections to one of four sets of
tubesheets 27. Tube 28 to tubesheet 27 connections of FIGS. 6A-6D are
illustrative, not all tube 28 to tubesheet 27 connections are shown, for
example tubes 28 may operatively connect to any or all portions of the
tubesheet 27.

[0038]The helical steam generator 20 is operatively supported by the upper
vessel 3. In one suitable support arrangement, the inside diameter is
increased at an upper end of the upper vessel 3 to provide a seating
surface 36 to support the steam generator 20, and the weight of the steam
generator 20 is supported from this point via the upper steam generator
support structure 35. In a suitable illustrative embodiment, the upper
vessel is composed of multiple ring forgings of SA508 Gr 4N Cl 2 carbon
steel, and the final weldment is clad internally with 0.25'' stainless
steel.

[0039]With reference to FIG. 1 and with further reference to FIGS. 7A and
713, a upper vessel head 4 is show as separate PWR component.
Alternatively, the upper vessel head may be integral with the upper
vessel 3, in which case the steam generator 20 and upper shroud 24 are
optionally supported from lugs on the inside of the upper vessel head 4.
The upper vessel head 4 suitably includes attachment nozzles 7 for
in-core instruments, the reactor coolant inventory and purification
system (RCIPS) spray nozzle and a vent nozzle. The upper vessel head 4
may optionally utilize either more of less nozzles 7 than that
graphically illustrated in FIG. 7A. The upper vessel head 4 may
optionally also includes lifting lugs 8 capable of lifting both the upper
vessel 3 and head 4.

[0040]With continuing reference to FIGS. 1 and 5 and with further
reference to FIGS. 6A-6D, some illustrative steam generator 20
embodiments are disclosed. Steam generators 20 transfer heat from the
primary coolant to the secondary feedwater in order to generate steam for
driving the turbine-generator. In the disclosed integral reactor design
the primary reactor coolant flows across the outside of the tubes 28 and
the secondary coolant flows inside the tubes 28, and the reactor pressure
vessel 10 contains the reactor core 50, steam generators 20 and primary
cooling system. The reactivity control rod drive mechanisms 40 are also
optionally wholly contained in the pressure vessel 10.

[0041]In some suitable embodiments that optimize the reactor vessel space
available for steam generation, a helical-coil, once-through steam
generator is used in integral reactor vessel. This steam generator
concept is depicted in FIGS. 6A-6D and 8-13 and in a suitable embodiment
uses 726 Inconel tubes of 0.75-inch outside diameter tightly wrapped in
an annulus created by the central primary flow riser pipe above the core
50 and the inside diameter of the reactor vessel. The aforementioned
Inconel tubes are merely an illustrative example, and other tubing
materials and sizes are also contemplated. The helical tubes 28 offer
some specific advantages from the standpoint of heat transfer. The
helically-coiled tubes 28 are arranged such that liquid flow, or
feedwater, enters the tubes above top of the reactor near the mid-level
of the vessel at a feedwater inlet 21, and spirals upward as it absorbs
energy from the downward flowing reactor coolant outside of the tubes in
the steam generator annulus. The liquid inside the tubing is at lower
pressure than the reactor coolant fluid, and transitions to steam along
the length of the steam generator. When the fluid in the tube reaches the
steam generator outlet 22, it has transitioned to pure steam, at a
temperature approximately 50° F. above its saturation point. This
provides steam delivered to the power-generating steam turbine that is
free or substantially free of undesired liquid droplets.

[0042]The designed helical-coil steam generator is effectively a
once-through, two phase, counter-current heat exchanger with boiling
occurring on the inside of the tubes. The secondary fluid enters the
bottom of the steam generator in a subcooled single phase condition, and
is heated by forced convection to the point of saturation. The saturation
point generally occurs in the lower 20% of the tube bundle length. In
practice, subcooled boiling may occur prior to the point of bulk fluid
saturation. In the following, however, the effects of subcooled boiling
on heat transfer are conservatively ignored. After the point of
saturation, nucleate boiling is initiated in the tube. Nucleate boiling
provides high heat transfer rates, resulting in the rapid vaporization of
the fluid as it travels up the tube. As higher secondary fluid qualities
are reached, departure from nucleate boiling occurs, along with the
associated decrease in heat transfer rates. At the upper end of the
generator, the fluid is completely vaporized, and heat is transferred to
the steam via forced convection. The primary fluid from the reactor core
enters at the top of the steam generator, and traverses down outside the
tubes or "shell side" of the steam generator. Heat transfer on the
primary side is due to single-phase, forced convection. Since the helix
angles are small (i.e. less than ten degrees) the flow is essentially in
a cross-flow orientation.

[0043]With reference to FIG. 8, a unique feature of two-phase counter-flow
exchangers when compared to single-phase exchangers is the existence of a
thermal "pinch point" that can limit heat transfer performance. FIG. 8
shows the temperature profiles for a hypothetical case on the primary and
secondary sides of the generator. The secondary fluid quickly heats up in
the forced convection region at the bottom of the bundle. As vaporization
initiates, the secondary temperature profile flattens out until the
superheated steam region, where temperatures again rise. Depending on the
primary flow rate and entering temperature, the secondary fluid
temperature may closely approach that of the primary at the end of the
forced convection region. This reduces heat transfer to the secondary
side, and greatly increases the amount of surface area needed to satisfy
the heat transfer requirements. This undesired effect on heat transfer
can be alleviated, for example, by increasing the primary fluid
temperature and/or flow rate, or reducing the secondary delivery
pressure.

[0044]FIGS. 9, 10, and 11 illustrate the results of a parametric analyses
relating to steam generator design. For a given bundle geometry, the tube
length (and hence the overall bundle coil heights) was varied until the
desired heat transfer load was achieved. FIG. 9 for example, shows the
parametric results for a six-degree helix. Combinations of bundle height
and reactor vessel diameter which satisfy the overall heat load design
parameters are plotted, with the steam generator exit pressure as a
parameter. Secondary exit pressures from 300 to 600 psia were analyzed.
FIGS. 10 and 11 show similar results for an eight and ten-degree helix
angle, respectively. All three helix angle cases illustrate the trade-off
between bundle height and vessel diameter which exists for a given heat
load requirement. As the vessel diameter is decreased, the bundle length
required for heat transfer increases. Higher exit pressures require a
longer bundle for a given reactor diameter. A comparison of FIGS. 9-11
shows that for a given reactor diameter and exit pressure, the required
bundle height goes down as helix angle is increased. This results from
the increased number of tubes and the associated thermal-hydraulic
changes on the secondary side. The increased helix angle therefore allows
for smaller vessel diameters to be considered, while maintaining the
overall length within reasonable limits.

[0045]FIGS. 9-11 further illustrate that an increase in helix angle
results in more tubes and a lower secondary pressure drop for a given
space envelope. The higher angle also results in a shorter helical
bundle. However, a low helix angle insures that the shell-side flow
remains essentially in pure cross-flow, thus maintaining high shell-side
heat transfer coefficients, and the slanted orientation of the tube, in
addition to the coiled flow path, result in a smoother transition from
nucleate boiling to film boiling, unlike vertical tubes, which tend to
feature a more abrupt transition. Although analysis of helix angles
greater than ten degrees are not shown, helix angles of greater than ten
degrees, including up to 15 degree, are contemplated for use in the
disclosed helical steam generators. Similarly, helix angles of less than
six degrees, including down to about 4 degrees, are contemplated for use
in the disclosed helical steam generators

[0046]Based on the results of the parametric studies, an eight-degree
helix angle, with a 120.0-inch vessel inside diameter, and a helical
bundle height of 28.9 feet (corresponding to a tube length of 208 feet)
was selected as a baseline case.

[0047]FIGS. 12 and 13 further illustrate a steam generator design of an
integral steam generator which includes an array of helically-coiled
tubes placed in the annular region between the outside diameter of the
upper shroud and the inside diameter of the upper shell. The tubes are
arrayed in coil rows in which the number of tubes and the helix radius
increase progressively in going from the innermost to outermost coil row.
The number of tubes in each coil row progresses approximately linearly
with the increase in coil row radius. Each tube in a particular coil row
exhibits the same helix angle, however the helix angle varies slightly
from coil row to coil row, being distributed around the overall bundle
nominal helix angle. The variation in helix angle from coil row to coil
row comports with the tubes in a coil row being equally distributed
circumferentially, and with the number of tubes in the set being an
integer. The lateral pitch is defined as the radial distance between coil
row centerlines. The vertical pitch is defined as the vertical distance
between tube centerlines. Both of these pitch values remain constant at
all locations in the bundle, as diagrammatically shown in FIG. 13.

[0048]The steam generator tubes enter the vessel at the lower end of the
steam generator section, through a number of tube sheet assemblies
originating around the feedwater inlet 21. The tubes are then routed from
the nozzle to the helical section of the steam generator, and progress in
a helical pattern up to the top of the steam generator section, finally
being routed to the exit tubesheet assemblies 27 shown graphically in
FIGS. 6A-6D. The tube runs extending from the nozzles to the helical
section of the bundle are referred to as "pigtails". If the pigtail
entrance and exit regions are ignored, each tube in the bundle has
approximately the same total length and axial height (slight variations
occur from coil row to coil row due to the variations in helix angle).
The number of turns varies inversely with radius from coil row to coil
row. For the illustrative example, the coil rows are assumed to be
co-wound, i.e. all coil rows are of the same helical direction. It is
noted that although the lateral and vertical pitch values remain constant
at each location in the bundle, the relative orientation of the adjacent
coil rows varies continuously with circumferential angle between in-line
and staggered.

[0049]A mechanical support for the helical steam generator tube array is
optimally composed of interlocking combs or support structure which
supports each tube at various circumferential locations. The support
assemblies may then be attached to bottom of the steam generator support
structure 35.

[0050]With returning reference to FIGS. 1, 2, 3, and 3A the DHRS is
described further. The DHRS provides a redundant method of removing core
decay heat in the event that the normal, non-safety, heat removal systems
are unavailable. The illustrated DHRS includes a plurality of independent
closed-loop heat removal systems that operate by natural circulation.
Each loop extends from a helical coil heat exchanger 55 at the top of the
lower reactor vessel, to one of two water pools located outside of the
containment. Water from the pools flows into the DHRS heat exchangers
where it is turned to steam. The steam flows back into the pool and is
discharged through spargers into the pool. In some embodiments, each DHRS
loop is capable of removing 1.8% to 2.4% of rated core power when the
reactor is at normal operating pressures and temperatures, thus enabling
any one of the plural (e.g., four) loops to remove 100% of decay heat
within approximately ten minutes after reactor shutdown (worst case) and
two loops remove 100% of decay heat within approximately 15 seconds after
reactor shutdown (worst case). The dedicated DHRS can operate over a
variety of temperatures allowing the system (all four loops) to remove
approximately 1.3% of rated core power when the reactor pressure is
reduced to 50 psia. With two of the four loops available, the system is
enabled to handle 100% of core decay heat within approximately six hours
after shutdown.

[0051]Because the dedicated DHRS heat exchangers are located in the lower
reactor vessel, the system can passively remove decay heat during all
phases of plant operation, including refueling. The DHRS utilizes four
heat exchangers inside the lower reactor vessel to remove heat from the
reactor coolant without allowing the reactor coolant to escape from the
reactor vessel.

[0052]In a suitable embodiment, each DHRS loop consists of approximately
14 tubes that are 1.5 inches in outside diameter. The four sets of heat
exchanger tubes are optionally wound together to form a single helical
bundle in the reactor vessel down corner. Each loop has separate inlet
and outlet tube sheets to assure complete isolation and the loop in some
embodiments is designed to 1600 psia up to the outside containment
isolation valve, allowing a loop with a tube leak to be isolated without
releasing reactor coolant inside or outside of the containment.

[0053]Two large, approximately 30,000 gallon, pools of water serve as
ultimate heat sink for the DHRS in the event of loss of normal heat
removal systems. During normal operation, the pools are cooled by the
plant service water cooling system which rejects plant waste heat through
the main cooling system towers.

[0054]The illustrative reactor of FIG. 1 is a natural circulation integral
reactor with reactor core 50, steam generators 20, control rod drives 40,
and decay heat removal heat exchangers 55 located inside a single reactor
vessel 10. The reactor pressure vessel 10 is divided into three sections.
The lower section 1 houses the reactor core 50. The core 50 is located
within a shroud separating riser and downcomer sections. Near the top of
the lower vessel 1, in the downcomer region, are four independent DHRS
heat exchangers 55 designed to remove core decay heat in the event that
normal heat removal paths are lost. Bypass flow holes 67 in the core
shroud above the steam generator allow coolant flow, even when the
reactor vessel is only partially filled with water. There are no reactor
coolant penetrations in this section. The upper portion of the lower
section 1 provides a support flange 53 for the upper internals 2. The
upper internals structure 2 is composed of an upper internals basket 46,
a CRDM support structure 48, control rod guide frames 44, a mid flange
29, and the control rod drive mechanisms 40. The upper vessel 3 houses
the steam generators 20. The upper vessel head 4 attaches to the top of
the upper vessel 3 and may optimally include a steam bubble generated by
core heat to pressurize the reactor. Both the upper vessel 3 and upper
spool 6 of the lower vessel 1 may be removed during refueling, allowing
the steam generator inspection to be conducted away from the reactor
core.

[0055]The illustrative integral PWR of FIGS. 1-7B employs natural
circulation, in which heating of the primary coolant by the reactor core
50 causes the primary coolant to circulate by flowing upward through the
riser 90 defined by the upper shroud and back downward in the outer
annulus 92 defined by the shroud and the pressure vessel. The downward
flowing primary coolant interacts with the steam generators.

[0056]With reference to FIG. 14, an alternative design is illustrated,
which employs forced convection powered by primary coolant circulating
pumps. The design of FIG. 14 is also an integral pressurized water
reactor (PWR) design, and has a three-section design including: (i) a
lower vessel 1 similar to that depicted in FIG. 2; (ii) upper internals 2
similar to those depicted in FIG. 3; (iii) an upper vessel 3 similar to
that depicted in FIG. 5; and (iv) an upper vessel head 4 similar to that
depicted in FIGS. 7A and 7B, which may or may not be integral with the
upper vessel 3. The forced convection design of FIG. 14 includes integral
steam generators, which in some embodiments are suitably embodied by the
helical steam generators described herein.

[0057]With reference to FIG. 4 the circulating pump feature of a forced
convection design is illustrated in FIG. 14. The mid-flange 29 supports
the upper internals including, for example, the control rods 40 and
internal control rod drive mechanism (CRDM) support structure (not
shown). The mid-flange 29 in the force convection reactor of FIG. 14 also
supports a plurality of primary coolant circulating pumps 85. The
circulating pumps 85 are wholly internal to the pressure vessel 10, for
example embodied as canned pumps designed to withstand the operating
temperature, pressure, and radiation fluence environment inside the
pressure vessel.

[0058]Advantageously, the circulating pumps 85 are located at the
mid-flange 29 in the annular region between the shroud and the pressure
vessel wall. This relatively central location facilitates even flow of
the primary coolant. The circulating pumps 85 are optionally not directly
coupled with the steam generators, which simplifies connections. The
circulating pumps 85 each comprise an impeller that is not connected with
any input or output piping.

[0059]It will be further appreciated that the octagonal reactor fuel core
support 51 illustrated in FIG. 2A for the natural convection reactor is
also suitable for use in the forced convection reactor. Moreover, the
disclosed octagonal reactor fuel core support can be replaced by other
polygonal configurations, such as a hexagonal reactor fuel core support,
where again gaps between the walls of the hexagonal reactor fuel core
support and the pressure vessel define spaces that can accommodate the
dedicated passive DHRS heat exchangers. Round and elliptical
configurations are also contemplated.